### RERTR Publications:

Foreign Research Reactor Spent Nuclear Fuel

#### ANL/RERTR/TM-26

##### APPENDIX C

MASS INVENTORY ESTIMATE: ORIGEN VS. WIMS

In this paper, the spent fuel nuclear mass inventories are based upon material number densities calculated within the WIMS code using burnup dependent cross sections and fluxes to solve the material transmutation equations. Unit-cell models of MTR, TRIGA and DIDO fuel assemblies with typical fuel compositions used WIMS to generate actinide cross sections and number densities as a function of U-235 burnup.

Spot checks of the mass inventories for two TRIGA fuel compositions were also calculated using the isotope generation and depletion code, ORIGEN. The principal actinide cross sections input to ORIGEN were collapsed one-group, zero-burnup material cross sections calculated by WIMS.

The fuel material mass inventories predicted by ORIGEN and by WIMS for TRIGA
fuel materials (133 g HEU and 38 g LEU) with 35% U-235 burnup are shown in Table
C1. The uranium isotopes at the 1-gram level and the Np, Pu and Am isotopes
at the 0.1-gram level are in reasonably good agreement. The slightly larger
^{237}Np and ^{239}Pu and the slightly smaller ^{238}U
inventories are due to the use of zero-burnup cross sections in estimating the
35% U-235 burnup inventories.

**Table C1. Gram-Mass Inventory Estimates**

8.5wt% U, 70% Enrichment
133 g U-235 |
8.5wt% U, 20% Enrichment
38 g U-235 |
|||

ORIGEN | WIMS | ORIGEN | WIMS | |

U-235 Burnup, % | 35 | 35 | 35 | 35 |

U-235 Burned, g | 47 | 47 | 13 | 13 |

U-234 | 0 | 0 | 0 | 0 |

U-235 | 86 | 86 | 25 | 25 |

U-236 | 8 | 8 | 2 | 2 |

U-238 | 55 | 55 | 150 | 151 |

U | 150 | 150 | 177 | 177 |

Np-237 | 0.4 | 0.3 | 0.0 | 0.0 |

Np | 0.4 | 0.3 | 0.0 | 0.0 |

Pu-238 | 0.0 | 0.0 | 0.0 | 0.0 |

Pu-239 | 1.3 | 1.1 | 0.9 | 0.9 |

Pu-240 | 0.2 | 0.2 | 0.1 | 0.1 |

Pu-241 | 0.1 | 0.1 | 0.0 | 0.0 |

Pu-242 | 0.0 | 0.0 | 0.0 | 0.0 |

Pu | 1.6 | 1.4 | 1.1 | 1.0 |

Am-241 | 0.0 | 0.0 | 0.0 | 0.0 |

Am | 0.0 | 0.0 | 0.0 | 0.0 |

The mass inventories using the WIMS 35%-burnup material cross sections as
input to ORIGEN shows a change in the inventories in the direction of the WIMS
results. In particular, the ^{237}Np and ^{239}Pu inventories
decrease and the ^{238}U inventory increases. With these cross sections,
the ^{237}Np and ^{239}Pu inventories are slightly underestimated
compared to WIMS. The cross sections used in ORIGEN are not extremely sensitive
to burnup, but they should be for a specific fuel material composition and not
simply default library cross sections. The difference between ORIGEN and WIMS
inventories would be expected to increase as U-235 burnup increases.

Since the Np and Pu mass inventories calculated above are slightly overestimated using zero-burnup cross sections and slightly underestimated using 35%-burnup cross sections, it is recommended that mid-cycle burnup cross sections be used in any ORIGEN mass inventory calculation. The mid-cycle cross sections would be expected to approximately cancel any over- or under-estimate and give inventory masses of U, Np, Pu and Am closer to the masses calculated with WIMS.